At present, stainless steel JIS 304 or 316 has been used for structural members for a core structure and equipment of a light water reactor, and measures have been taken for such steel to enhance a corrosion resistance and a resistance to intergranular fracture susceptibility in an environment in which such steel is used. For example, as a method of relieving or preventing intergranular cracking from occurring due to intergranular corrosion and intergranular stress corrosion cracking (hereinafter referred collectively to as "intergranular corrosion") in steel used in high-temperature water, there is known a method, as disclosed in Japanese Patent Examined Publication No. 1-18143 and Japanese Patent Unexamined Publication Nos. 62-238355 and 62-238353, in which the amount of carbon contained in stainless steel is reduced, and a carbide stabilizing element such as Nb and Ti is added to suppress or prevent a Cr exhaustion phenomenon from occurring in the vicinity of a grain boundary. Japanese Patent Unexamined Publication No. 62-107047 discloses a method of preventing intergranular corrosion from occurring due to neutron irradiation, in which Si and P are reduced, and an element for producing stable carbide, such as Mo, Nb and Ti, is added to suppress the formation of carbide of Cr. Further, Japanese Patent Unexamined Publication No. 3-72054 discloses a method in which the arrangement of components is adjusted in such a manner that a volume size factor of Cr relative to an average atomic volume of a matrix phase in stainless steel is 0.900-1.030, thereby suppressing the mechanism of producing an irradiation-induced intergranular Cr-shortage layer. These techniques have been developed for the purpose of suppressing the occurrence of the Cr-shortage layer in the vicinity of the grain boundary to enhance a resistance to intergranular corrosion in high-temperature, high-pressure water, and have been intended particularly for stainless steel used for parts of a core structure of a nuclear reactor and a first wall of a nuclear fission reactor.
However, it has been found that when such stainless steel is provided in a core of a nuclear reactor, hydrogen, occurring or infiltrating in the stainless steel because of a high-temperature, high-pressure water environment and a neutron irradiation environment, is accumulated in the grain boundary to induce grain boundary embrittlement to increase intergranular fracture susceptibility. Hydrogen occurs or infiltrates in the stainless steel, used in a nuclear reactor, through the following process.
(1) (n, p) nuclear reaction of impurities in steel due to neutrons. PA1 (2) Decomposition (radiolysis) of reactor water due to .gamma.-ray. PA1 (3) Corrosion reaction due to high-temperature, high-pressure water.
When the hydrogen in the stainless steel is accumulated in the grain boundary, so-called hydrogen-induced grain boundary embrittlement occurs to increase intergranular fracture susceptibility of the member, thereby adversely affecting the resistance to neutron irradiation embrittlement.
In the above prior art techniques, no consideration has been given to the increase in the intergranular fracture susceptibility due to the hydrogen grain boundary embrittlement in the stainless steel, and they have a problem on resistance to intergranular cracking in high-temperature, high-pressure water under neutron irradiation.